SPENT FUEL STORAGE AND ITS TRANSPORTATION

The spent fuel storage facilities are designed, constructed and operated by following the international standards and safety guidelines. The spent fuels from the reactors are stored initially at the reactor pool. Since natural uranium is used as fuel in PHWRs, no specific configuration is required for fuel storage. The spent fuel from LWRs is stored in racks made of special stainless steel. The spent fuel is kept in the storage locations either at reactor site or away from the reactor site depending on the management strategy. In rectors under safeguard, the spent fuel is stored in reactor pools in compliance with the IAEA guidelines. Storage capacities at reactor site generally cater to store spent fuel for nearly 10 reactor-years of operation in case of PHWR fuels whereas it is about 5–7 years for LWRs. Dry storage of spent fuel is also adopted but to a much lesser extent.

Transport of the spent fuel to the storage locations are carried out adhering to IAEA safety guidelines following three lines of defence viz. the transport reliability, concept of a package and the efficacy of resources to deal with an accident. Spent fuel transport is carried out in ‘type B’ packages, designed to withstand severe accident conditions, simulated by tests, validated by approval certificates and subject to inspection. During transportation, security and safety issues are given topmost priority wherein physical security of nuclear materials from thefts, diversion etc. are covered besides exposure, contamination, criticality and environment related issues.

4. MANAGEMENT OF DAMAGED AND PREMATURE SPENT FUEL

Depending upon the degree of damage, such fuel elements are given special treatment. Fuel bundles are encapsulated safely and transported to the storage pool along with normal spent fuel for reprocessing or subjected to dry storage in special casks. The damaged fuel bundles are subjected to post irradiation examination by visual and using several non-destructive examinations viz. ultrasonic testing, eddy current testing, gamma scanning and

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neutron radiography for root cause analysis. Subsequently corrective measures are studied and implemented to minimize such damages to the fuels. Premature fuel is managed as normal spent fuel by reprocessing in special campaigns.

5. REPROCESSING

PUREX process has been the main workhorse of spent fuel reprocessing for the last several decades. The process utilizes 30% TBP in n-dodecane as solvent to extract uranium and plutonium from dissolved feed solution retaining bulk of fission products in the raffinate phase. Main steps involved are given below:

— Head-end treatment involving mechanical chopping of the spent fuel followed by dissolution in nitric acid.

— Feed clarification and conditioning of the feed solution for solvent extraction.

— Co-decontamination involving extraction of uranium and plutonium leaving bulk of the fission products.

— In the raffinate phase.

 Washing/scrubbing of organic stream with nitric acid;

 Reductive partitioning of plutonium from uranium using uranium nitrate solution;

 Purification of uranium and plutonium streams;

 Conversion of Pu as PuO2 via oxalate route and uranium as U3O8 via ADU route;

 Solvent wash and its recycle;

 Waste management.

In spite of numerous advantages, TBP is found to have certain disadvantages like higher aqueous phase solubility, formation of harmful degradation products like dibutyl phosphate, third phase formation in presence of higher concentration of tetravalent metal ions viz. Pu⁴⁺, Th⁴⁺ and Zr⁴⁺, non-incinerable nature and waste management issues. Based on extensive laboratory scale studies a C, H, O and N based monoamide viz. dihexyl octanamide was proposed and investigated under simulated feed conditions [2, 3]. Extraction and stripping behaviour is found almost similar to that of TBP under identical conditions. However, it is not yet fully established for reprocessing applications using real feed solutions. Behaviour of aqueous soluble degradation product is yet to be studied. It is also noted that hydrodynamic parameters like viscosity, frothing nature etc. are also not found to be very favourable.

Even though higher homologues of TBP such as tri-n-hexyl phosphate, tri amyl and tri-iso-amyl phosphate are extensively studied [4, 5] at IGCAR for processing spent fuel from fast breeder reactors, so far only TBP is deployed for reprocessing technology in India. As of now it is a challenge for any other solvent that can replace TBP in PUREX technology.

Though India has mastered the reprocessing technology to meet the present-day requirements and future challenges, the technology is constantly being improved. Several developmental activities are being pursued to enhance the process performance. Some of the areas where R&D is being continuously pursued are:

— Improvements in the recoveries of U and Pu;

— Improved decontamination factors for U and Pu with respect to fission products;

— Improvements in partitioning techniques;

— Reduction in waste volume generation;

— Reduction in number of cycles;

— Use of non-proliferation route following co-processing and co-conversion;

— Partitioning of useful actinide and fission products for societal application.

6. WASTE MANAGEMENT

India considers HLLW as a resource rather than waste. Separation and purification of several radionuclides from HLLW is being explored for various societal applications. Solvent extraction-based engineering facilities have been deployed to partition the waste for separation of active components like U, Cs-137, Sr-90 and An-Ln using indigenously synthesized organic solvents viz. TBP, calix crown and TEHDGA in three cycles.

59 Cycle I: In this cycle, Sulphate Bearing High Level Liquid Waste (SB-HLLW) is first contacted with 30%

TBP for recovery of uranium and plutonium. The stripped product i.e. U-rich solution from this cycle is sent back to the reprocessing plant for reuse. The lean organic is recycled.

Cycle II: After adjustment of feed acidity, the raffinate from first cycle is contacted with caesium selective calix crown 6 (CC6) solutions to extract Cs. Cs loaded in the organic phase is back extracted with water. The aqueous phase containing Cs-137 is concentrated and used to make vitrified glass pencils. Over 300 000 Ci of radio Cs has been recovered successfully so far from HLLW and used for vitrified glass pencils having specific activity of 2-5Ci/g. The vitrified pencils are supplied to BRIT, India which are being deployed for blood irradiation applications. Photograph of a typical vitrified Cs pencil is given in Fig. 1.

FIG. 1. Typical Vitrified Cs Glass Pencil.

Cycle III: The raffinate from the second cycle i.e. Cs and U lean HLLW is subjected to solvent extraction using TEHDGA, for separation/recovery of Sr and actinides. The stripped product from this cycle is predominantly rich in Sr-90 and contains actinides and lanthanides. This aqueous phase is concentrated and subsequently used as source of radio-strontium/vitrified.

The final raffinate generated from third cycle is subjected to interim storage to allow decay of short lived radio-nuclides if any and is further processed as low level waste prior to its discharge meeting the regulatory requirements. Process flow scheme deployed for the treatment of SB-HLLW at Trombay is shown in Fig. 2.

FIG. 2. Management of HLLW by Solvent Extraction Route.

The solvent extraction technology thus adopted for management of HLLW has significantly reduced the waste volumes in terms of vitrified mass and has extended the period for repository requirements.

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7. SEPARATION OF Sr-90 FOR GENERATION OF CARRIER-FREE Y-90 FOR THERAPEUTIC APPLICATIONS

For achieving radiopharmaceuticals grade purity of Sr-90, multi-step separation process involving solvent extraction, ion exchange, extraction chromatography, precipitation and membrane-based techniques are deployed.

Sr-90 rich stream generated from third cycle is used as feed. Sr-selective crown ether, di-(t-butyl cyclohexano)-18-Crown-6, is being synthesized and will be deployed for large scale separation of Sr-90. A two-stage SLM based generator system [6] (Fig. 3) is used for the separation of carrier free Y-90 which is principally based on the solvent extraction properties of two ligands, namely 2-ethylhexyl 2-ethylhexyl phosphonic acid (KSM-17) and octyl phenyl-N,N-diisobutylcarbamoyl methyl phosphine oxide (CMPO) under optimum conditions.

The carrier-free Y-90 acetate product lots having specific activity in the range 30-40Ci/L are supplied for Radiopharmaceutical application. To meet the higher Y-90 activity demand multiple such generator systems are under consideration.

Stage-1 Stage-2

FIG. 3. SLM based two stages for generation of carrier-free Y-90.

8. PARTITIONING OF ACTINIDES FROM PHWR-HLLW

An Actinide Separation Demonstration Facility was set up at Tarapur for the partitioning of actinides from PHWR-HLLW [7]. Block diagram of the integrated facilities is given in Fig. 4. In this facility, residual U and Pu are separated from the waste using PUREX solvent in first step. In the second step, actinides and lanthanides are partitioned from the waste using TEHDGA based solvent extraction process. Separation of actinides from lanthanides is proposed to be carried out using D2EHPA based process as adopted in TALSPEAK process [8] in the third step.